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; ; ; Daigo, Yoshimichi
PNC TN941 81-05, 235 Pages, 1981/01
Failure propagation tests have been carried out using the Steam Generator Safety Test Facility (SWAT-3) in PNC O-arai Engineering Center to establish the method of safety design of the LMFBR Monju's prototype steam generator with reference to preventing sodium-water reaction accidents. The main object of these tests is to understand how the failure propagation to heat transfer tubes around progresses owing to the water leakage from the initial nozzle. Here reported are the data of the first three failure propagation tests conducted in october 1979. Three injection tests (SWAT-3 Run-8 through 10) were executed whose initial water leak rates were 36 g/s, 6.8 g/s and 570 g/s respectively. (1)Failure propagation progressed in about each one minute in Run-8 and 10, but it did not occur in Run-9. (2)The maximum size of penetration holes is about 5.7 mm
for water tubes, and is 18 mm 
33 mm for gas tubes. (3)The main mechanism of failure seemed to be the wastage. (4)There were some bowing and buldging tubes as well as wastaged tubes in Run-8 and 10. (5)The wastage rate was less than 710
mm/s in accordance with results of intermediate wastage tests.
; ; ; ; ;
PNC TN941 80-27, 272 Pages, 1980/02
A series of intermediate leak wastage tests were conducted using Large Leak Sodium-Water Reaction Test Rig (SWAT-1) at O-arai Engineering Center. The purpose of these tests was to clarify the mechanism of failure propagation to adjacent heat transfer tubes due to the flame jet of leak water in the steam generator of LMFBR. Then the value of the design basis leak for the steam generator should be proposed. The wastage rate, the size of secondary failure, and the multi-tube wastage phenomenon were mainly discussed on the basis of eleven test results, in which the water leak rate was in the range of 10
200 g/sec. Findings are as follows ; (1)The wastage rate depends on L/D (L: nozzle to target distance, D : nozzle diameter) and has a maximum value of 7 10 mm/sec at L/D=2030. (2)The time of secondary failure occurring in the tube bunk structure does not depend on the water leak rate. (3)The maximum diameter of the penetration hole obtained in these tests was 19mm. (4)The dominant mechanism of the failure propagation is not overheating but wastage in the intermediate leak region. (5)The extent of multi-tube wastage increases with the increasing leak rate. Six tubes were damaged considerably at the leak rate of 200 g/sec.
Hiroi, Hiroshi*; ; ; ; ;
PNC TN941 79-04, 274 Pages, 1979/10
Large leak sodium-water reaction tests have been carried out using SWAT-3 facility in PNC O-arai Engineering Center to obtain the data on the safe design of the prototype, LMFBR Monju is steam generator against large leak accident. This report gives the results of SWAT-3 run-5 test. The heat transfer tube bundle of the evaporator used in run-5 test was designed and fabricated by MITSUBISHI HEAVY INDUSTRIES, LTD. The water injection rate into the evaporator was 15 kg/sec, which corresponds to test scale of 5 tubes failure in actural size system according to iso-velocity modeling. Measurements were made of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 24.6 kg/cm
a nearest to injection point, and the maximun quasi-steady pressure in evaporator was 7.6 kg/cma. The rupture disc of evaporator was bursted at 0.23 sec after water injected, and the pressure relief system was well functioned. No secondary tube failure was observed.
; ; ; Daigo, Yoshimichi;
PNC TN941 79-141, 198 Pages, 1979/09
When a large leak sodium-water reaction accident occurs in a steam generator of LMFBR, pressure relief piping might be received thermal shock or blocked by frozen sodium, if it is not preheated. Then, the thermal transient tests were performed using the large leak sodium water reaction test rig SWAT-1. The results are summarized as follows; (1)Four tests were executed. The water injection rate of two tests was equivalent to that of several DEG (double-ended guilotine) failure of heat transfer tubes considering the difference of evapourator inner diameters between "Monju" and SWAT-1, and in other two tests the injection ratio was equivalant to less than that of 1 DEG. (2)Flow pattern in the pressure relief piping of two large injection rate tests was as follows, void fraction was as low as that of sodium single-phase flow in its early stage of 0.20.3 sec., and rapidly increased to about 0.9. In case of the small injection rate tests, the stratified flow had continued for 23 sec., it was followed by hydrogen gas single-phase flow. (3)In the large injection rate tests the maximum value of heat flux was about 110
[kcal/(mh)], and that of heat transfer coefficient was 310
[kcal/(mh
C)] except in its very initial stage. In case of the small tests, they were lower. (4)In the large injection rate tests, stain of outer piping surface was about 8001,50010
, which agrees with the calculation using above value as heat transfer coefficient. (5)Possibility of blockage by frozen sodium is seemed to be very little in SWAT-1 test rig.
Hiroi, Hiroshi*; ; ; ; ;
PNC TN941 79-118, 282 Pages, 1979/06
Large leak sodium-water reaction tests have been carried out using the SWAT-3 facility in PNC O-arai Engineering Center to obtain data on the safe design of the prototype LMFBR Monju's steam generator with reference to preventing large leak accident. This report gives the results of SWAT-3 run-4 test. The heat transfer tube bundle of the evaporator used in Run-4 test was designed and manufaetured by TOSHIBA/IHI. Main purpose of this test is to clarify sodium-water reaction phenomena occured in the upper coil region, that is, the place near by sodium surface. Water was injected into the evaporator at the rate of 9.0 kg/sec, which corresponds to a test scale of 5 tube failure in an actural size system according to iso-velocity modeling. Measurements were taken of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 14.7 kg/cma closest to the injection point, and the maximum quasi-steady pressure in the evaporator was 5.4 kg/cma. The rupture disc of the evaporator burst 0.536 sec. after water was injected, and the pressure relief system functioned well. No secondary tube failure was observed.
; ; Hiroi, Hiroshi*; ; ;
PNC TN941 78-154, 210 Pages, 1978/10
Large Leak sodium-water reaction tests have been carried out using SWAT-3 facility in PNC O-arai Engineering Center to obtain the data on the safe design of the prototype LMFBR Monju's steam generator against large leak accident. This report describes the resulting data of Run-6 test. The heat transfer tube bundle of the evaporator used was fabricated by HITACHI/BABCOCK HITACHI. The water injection rate into the evaporator was 9.4 kg/sec, which corresponds to test scale of 5.7 tubes failure in actual size system according to iso-velocity modeling. Pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, etc were measured during water injection test. Initial spike pressure was 12 kg/cma nearest to injection point, and the maximun quasi-steady pressure in evaporator was 5.1 kg/cma. The rupture disc of evaporator was bursted at 0.58 sec after water injected, and the pressure relief system was well functioned. No secondary tube failure was observed.
Hiroi, Hiroshi*; ; ; ; ;
PNC TN941 78-93, 192 Pages, 1978/01
Large Leak sodium-water reaction tests have been carried out using SWAT-1 rig and SWAT-3 facility in PNC O-arai Engineering Center to obtain the data on the safe design of the prototype LMFBR Monju's steam generator against large leak accident. This report gives the results of SWAT-3 run-3 test. In run-3 test, the heat transfer tube bundle of the evaporator, fabricated by TOSHIBA/IHI, were used, and the pressure relief line was located at the side of evaporator. The water injection rate to the evaporator was 8.8 kg/sec, which corresponds to test scale of 3.5 tubes failure in actual size system according to iso-velocity modeling. Measurements were made of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 16 kg/cma nearest to injection point, and the maximun quasi-steady pressure in evaporator was 5.0 kg/cma. The rupture disc of evaporator was bursted at 0.56 sec after water injected, and the pressure relief system was well functioned. No secondary tube failure was observed.
; ; Hiroi, Hiroshi*; ;
PNC TN941 78-78, 33 Pages, 1978/01
The experimental study was performed to establish the safe design of the LMFBR MONJU steam generator system against large leak sodium/water reaction accident. Two tests were carried out to obtain the pressure relief system performance data using large leak sodium/water reaction test rig SWAT-1. In this report we tried to estimate the flow behaviour in the pressure relief pipes from data obtained by pressure transducers, gamma-ray densitometer, sodium void detectors etc. Water leak rates were chosen 0.7 and 1.4 kg/sec in these tests, and about 5 kg of water was injected into 180 kg of sodium in both. The results were summarized as follows. (1)Flow pattern in the pressure relief pipes was (divided into two, i.e. plug / bubbly flow at the initial stage and annular / drop-annular after this stage. (2)Though the initial term was too short, the removed sodium quantity during this term was about half of the total quantity of sodium expelled to the reaction product tank.
; ; ; ; ; ;
PNC TN941 78-32, 84 Pages, 1978/01
The study to establish the safe design of LMFBR MONJU's steam generator system against the large leak sodium-water reaction is conducing by Power Reactor & Nuclear Fuel Development Corporation. This report includes seven topics which have been presented to the meetings of Atomic Energy Society of Japan, etc. during 1977. Summaries of these topics are as follow; (1)A computer program SWAC-11 was developed to predict the water leak rate from the ruptured heat transfer tube in the steam generator. The basic equations used and method of numerical calculation were explained. The sensitivity survey calculations for various parameters used in the code and the demonstration calculation of the case of MONJU's evaporator were reported. (2)A computer program SWAC-13 was developed to ptedict the pressure and flow behavior in the secondary cooling system, i.e. the quasistatic pressure build up, sodium/hydrogen gas flow in the SG vessels and secondary sodium circuit, and flow characteristics of pressure relief line and the hydrogen gas reliese into the atmosphere. The modeling, basic equations and the method of numerical calculation of the program were reported, and the applicability to SWAT-3 tests was demonstrated. (3)The reaction vessel drain line of SWAT-3 facility was chocked with sodium-water reaction products after water injection test Run-3. In order to understand the cause of chocking, the investigations of the reaction product and temperature distributions in the drain line, and the chemical analysis and the freezing temperature measurement of the reaction products were performed. Those results were summarized quantitatively in this chapter. (4)In order to make clear the flow characteristic of the pressure relief line, we tried to rearrange the data of two test runs of SWAT-1 rig which were obtained by the various kind of the sensors attached on the relief line. It became clear from those investigations that the sodium/hydrogen-gas two phase flow in the ...
; ; ; Y.W.Shi*
PNC TN952 77-05, 87 Pages, 1977/05
For the analysis of large scale sodium/water reaction accidents in LMFBR steam generators, a computor program, SWAC-11, is developed to predict water injection from double-ended guillotine rupture of heat transfer tubes. Considering the phase change in water flow, SWAC-11 can simulate the injection of subcooled water, saturated water and superheated steam. The method of numerical calculation in the code is the modified ICE (Implicit Continuous Eulerian) Method for the thermal-hydraulic transient in heat transfer tubes, while the volume-junction model is employed for the other part of water system. Simplified problems were selected to make a fundamental check of the model. Moreover, to show the applicability of the SWAC-11 to the prototypic conditions, the water/steam leak from four tube double-ended guillotione rupture in the MONJU class evaporator was calculated as an example, This code is written in JIS-FORTRAN computor language and requires 120 K bytes core memories. Typical running time for a problem with 76 cells in a tube is about 70 minutes for FACOM-230/58 to compute 5,000 time steps.
; ; ; ; ; ;
PNC TN941 76-84, 63 Pages, 1976/08
PNC and CRIEPI have jointly perfomed a simulation experiment for pressure wave propagation in LMFBR steam generators during sodium/water reaction. The purpose of this work was to contribute to the safe design of the steam generators and from the results of this project the following has been established: (1)Analytical method has been devised to obtain the equivalent cross-sectional area for a system with inner structures, and (2) Structural dynamic response of the shell has been measured for given pressure wave form. By the simulation experiments, the equivalent cross-sectional areas have been established for Monju SG in its various design alternatives and, moreover, verification was made of the equivalent cross-sectional area for straight pipe intervals. In item (2) above, dynamic strain of the vessel wall was measured for the given simulated pressure pulse.
